ML20202C366
| ML20202C366 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 11/18/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20202C362 | List: |
| References | |
| NUDOCS 9712030318 | |
| Download: ML20202C366 (6) | |
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}AFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION
.I RELATED TO AMENDMENT NO. 223 TO FACILITY OPFRATING LICENSE N0. DPR-53 I
AND ANENDNENT No. 199 TO FACILITY OPERATING LICENSE NO. DPR-69 RgJIMDRE GAS AND ELECTRIC CONPANY CALVERT CLIFFS NUCLEAR POWER PLANT. UNIT NOS. 1 AND 2 DOCKET NOS. 50-317 AND 50-318 r
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1.0 INTRODUCTle'..
By letters dated November 30, 1995, March 15, 1996. March 6, 1997, and June 27, 1997, Baltimore Gas and Electric Company (BGE), the licensee, submitted a request to change the Technical Specifications (TSs) at Calvert Cliffs Nuclear Power Plant, Unit Nos. I and 2.
The proposed changes would incorporate references to a new Combustion Engine 9 ring, Inc. topical report describing steam generator tube sleeves, delete r=ferences to the previous CE topical re criteria, port, incorporate sleeve / tube inspection scope and expansion revise the plugging limit for a CE sleeve to 28% of the nominal -
sleeve wall thickness, and incorporate a post weld heat treatmat of free span welds.
The updated CE Topical Report, CEN-630-P, Rev. #01, ' Repair of 3/4" 0.D. Steam i
Generator Tubes Using Leak Tight Sleeves," dated November 1996, addresses issues identified in 1996 at Prairie Island Unit 1.
At issue, were indications detected in weld joints of CE sleeves resulting from inadequate cleaning.
Because the bulk o' the technical and regulatory issues for the present request are identical to those reviewed in the previous Safety Evaluations (SEs), this SE discusses only those issues warranting revision, amplification or inclusion based on current experience.
Details of prior staff evaluations of CE sleeves may kiso be found in the SEs for Waterford Steam Electric Station, Unit 3, Docket No. 50-382, dated December 14, 1995; Byron Nuclear Power Station, Units 1 and 2 and Braidwood Nuclear Power Station, Units I and 2 Docket Nos. 50-454, 50-455, 50-456, and 50-457, dated April 12, 1996. These evaluations relate to the proposed-Calvert Cliffs license amendment. The March 15, 1996, March 6, 1997, and June 27. 1997, letters provided clarifying infomation that did not change the initial proposed no significant hazards consideration determination.
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2.0 BACKGROUND
AND SupMARY OF PREVIOUS REVIEWS The proposed CE sleeve types are an expansion transition zone (ETZ) sleeve and a tu>e support sleeve. An ETZ sleeve is designed to restore the portion of a 4
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s tube in the vicinity of the top of the SG tubesheet. A tube support sleeve can be used to span a support plate elevation or be used on a freespan section of tube.
The CE sleeves are installed using gas tungsten are welding to join the sleeve to the parent tube at the upper (free span) end of the ETZ sleeve and at both ends of a tube support sleeve. The weld joint is the subject of the modifications to the instellation processes described in the new topical report. The lower ETZ sleeve tube joint is hard-rolled into the tubesheet below the expansion zone.
Thert are no changes from the previous topical reports with respect to the rolled joint.
The sleeve material is a nickel-iron-chromium alloy, alloy 690, a Code approved material (ASME SB-163), incorporated in ASME Code Case N-20.
This material has been extensively tested by a variety of investigators and found to be greatly superier in corrorion resistance compared to the existing SG tube material, alloy 600.
As detailed in the generic topical, the structural integrity of the sleeve design has been investigated analytically and verified by laboratory tests of sleeve mock-ups. The sleeve design meets all applicable ASME Code and Regulatory Guide 1.121 requirements.
Leak resistance of the rolled joint has been demonstrated through laboratory tests.
It is noted that the sleeve design is inherently leak limiting.
Bounding calculations and laboratory tests have verified that, should leakage develop, the 10 CFR Part 100 guidelines for radiological release would not be impacted, even under the most severe postulated conditions.
3.0 DISCUSSION Experience with all types of SG tube sleeves has revealed certain issues outside the scope of betic sleeve design and qualification discussed in previous SEs.
The principal topics are weld preparation and weld acceptance inspections.
During the spring 1996 refueling outage at the Prairie Island facility, roughly 60 previously installed weld joints in CE sleeved tubes were discovered to have eddy current test (ET) indications. This discovery was the result of the licensee employing a new, more sensitive ET probe for its periodic inspection of SG tubes, Tube / sleeve assemblies were removed from the SGs for metallurgical examination and root cause determination.
It was found that the ET indications were due to entrapped oxider and/or weld shrinkage within the sleeve to tube weld. The cause of these weld defects was traced to a previously revised tube cleaning procedure. Although the discovered weld defects did not significantly impair the structural integrity (strength) of the welds, they did pose a small leakage potential which is contrary to ASME Code and regulatory requirements for a welded joint.
As a result of the metallurgical examination, the tube cleaning procedure was revised and revised post cleaning visual inspections (VT) were adopted.
The initial weld acceptance inspection, an ultrasonic test (UT), was revised to
$ give greater sensitivity. As added measures, an optional VT of the completd welds was added to the installation procedure, and, the initial baseline ET, normally used only as reference for later periodic reinspection, was modified to supplement the UT as part of the initial weld acceptance inspection. All of these refinements to the sleeving procedure were confirmed using a large number of-laboratory samples and field mock-ups. These modifications were incorporated into a new generic Topical Report, CEN-630-P, referenced above and are discussed in more detail in the following sections.
3.1 Weld Preparation Prior to performing ary weld, the surface of the metal (s) to be welded must be cleaned.
For sleeve installation, the inner diameter of the parent tube at the desired weld location must be cleaned of service induced oxides.
For the CE sleeving process, this is accomplished using motorized wire brushes.
Based upon the metallurgical examination of the Prairie Island samples, CE
-revised the cleaning method to ensure optimum removal of service induced oxides. The revised cleaning procedure entailed some equipment changes. More significantly, from a quality assurance standpoint, a 100 percent VT of the cleaning process was instituted.
After the wire brush cleaning step, every tube is given a VT using a remote fiber optic camera system to confirm that adequate surface cleaning has been accomplished. CE advises that the 100 percent VT is an interim step until enough field experience is gained to consider adoption of a statistical sampling plan in the future..
3.2 Weld Acceptance Inspections For compliance with the Code and regulatory requirements, initial and periodic examinations of steam generator tubes and sleeves are performed. Sleeve welds were historically accepted based upon VT and UT examinations.
ET was used for an initial baseline inspection for comparison with later required periodic inspections. The reason for the different types of nondestructive examinations (NDE) being used for initial acceptance versus periodic reinspection is due to the differences between potential flaws arising from initial installation versus service induced degradation. The different NDE techniques have normally been better suited for the respective types of 4
- anticipated flaws.
The Prairie Island experience suggested that the then current initial acceptance examinations (VT and UT) may not be sufficient in every circumstance. As a result, the weld acceptance NDE was modified to include:
- 100 percent UT with an enhanced digitized amplitude system
- 100 percent ET using the Plus Point probe The original UT procedure was based upon the absence of a mid-wall reflection.
When fusion existed, the aid-wall reflection (mid-wall of the fused sleeve and tube combination) would not appear since no interface would exist.
The Prairie Island experience led CE to discover that lack of fusion caused by 4
o axially oriented oxide inclusions from an inadequately prepared surface would not be detected by the UT procedure in use. The oxides are sound conductors and do not cause a large sound reflection.
In the enhanced UT procedure, the back wall signal from the outside of the parent tube f5 also monitored for presence in the fused area. Additionally, the back wall signal strength is examined for excessive attenuation.
Attenuation beyond the normal amount can be interpreted, along with other signal artifacts, as either a weld that is to: aarrow or one with inclusions or patches of unfused material. The modified bT procedure was extensively tested on laboratory produced welds containing a variety of ir.clusion/ lack of fusion defects. Samples were then destructively examined and the metallurgical sections compared with the UT results. Comparison of results demonstrated that the revised UT procedure was highly reliable. No significant defects could remain undetected by the enhanced UT procedure.
ET with the Plus Point probe is now part of the sleuve weld acceptance criteria.
CE also discovered that weld shrinkage and circumferential1y oriented oxide inclusions from a poorly cManed weld would not be detected by UT.
CE has shown the Plus Point probe reliably detects these process-induced weld defects and blowholes. CE has also shown the ET can reliably locate the position of the defect with respect to the weld centerline which is considered the pressure boundary.
ET indications located above the weld centerline that meet UT requirements can be left in service. Any ET indication founo below the weld centerline requires the tube to be plugged.
The preparation, installation and initial inspection of CE welded sleeves, as outlined in sections 3.0. 3.1, and 3.2 above, are governed by the CE generic Topical Report CEN-630 i, Rev. #01 and is acceptable for use at Calvert Cliffs in that it will provide assurance of steam generator tube weld integrity.
By proposed modification to the plant TS 4.4.5.4a.10b, the generic topical would govern document for such CE sleeve installation. Additional licensee TS change proposals, beyond those of the CE Topical Report, are discussed below.
3.3 Sleeve Plugging Limits The sleeve minimum acceptable wall thickness is determined using the criteria of Regulatory Guide (RG) 1.121, " Bases for Plugging Degraded PWR Steam Generator Tubes" and ASME Code Section III allowable stress values and pressure stress equations. According to RG 1.121 criteria, an allowance for nondestructive evaluation (CE) uncertainty and postulated operational growth of tube wall degradation within the sleeve must be accounted for when using NDE to determine sleeve plugging limits. Therefore, a conservative tube wall combined allowance for postulated degradation growth and eddy current uncertainty of 20% throughwall per cycle was assumed for the purpose of determining the sleeve plugging limit. The sleeve plugging limit, which was calculated based on the most limiting of normal, upset, or faulted conditions for 3/4-inch outside diameter steam generator tubes in CE designed generators, was determined to be 28% of the sleeve nominal wall thickness based on ASME
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Code minimum material properties in accordance with staff positions. Removal of tubes and/or sleeves from service when degradation reaches a plugging limit of 28% provides assurance that the minimu,; acceptable wall thickness will not be violated during the next subsequent cycle of operation. Adoption of the 28% plugging limit is reflected in the proposed wording for TS 4.4.5.4.a.7.c.
- and is acceptable.
3.4 Post Weld Heat Treatment Accelerated corrosion tests confirm that a post weld heat treatment (PWHT) significantly improves the intergranular stress corrosion cracking resistance of the alloy 600 parent tube material in the weld zone.
In its June 27, 1997, submittal, the licensee committed to perfoming PWHT of the welded joints in accordance with'the CE generic sleeving report and the NRR staff position.
This commitment is reflected by proposed revision to plant TS 4.4.5.4.a.10.b.
and the TS chang 6 is acceptable.
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_- 3. 5 Adoption of Revised Inservice Inspection Expansion Criteria i
The licensee has proposed,_ through the incorporation of TS Table 4.4-3, to a minimum 20% (with expar: ion criteria) inservice inspection of all sleeved p
tubes in accordance with industry guidelines. The staff findt, this inspection sample acceptable.
3.6 Staff Conclusions The staff concludes the proposed sleeving repairs, as described in the new CE i
sleeve topical report, can be accomplished to produce sleeved tubes of
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acceptable structural integrity, leak tightness and corrosion resistance.
The staff also finds the proposed preservice inspection methods for examining the conditions of the welds are acceptable.
I The NRC staff concludes the repair of SG tubes using welded sleeves designed by CE.is acceptable, as discussed above and refiscted by.TS revisions, to:
1): install welded CE sleeves in accordance with CE Topical Report CEN-630-P, Rev. #01, " Repair of 3/4" 0.D. Steam Generator Tubes Using Leak Tight i
Sleeves," dated November 1996, 2) performing PWHT of the free span weld
_ joints, 3) incorporating TS table 4.4-3 for sleeve-inservice inspection sample size and expansion criteria, and, 4) revising the sleeve plugging limit to 28%
of-the sleeve wall. thickness. The bases have also been modified consistent-with'these TS changes.
4.0 STATE CONSULTATION
2 In-accordance with the Commission's regulations, the Maryland State official was notified of the proposed issuance of the amendments. The State official had no comments.
4 5.0 ENVIRONMENTAL. CONSIDERATION The' amendments change a requirement with respect to installation or use of a
-facility component located within the restricted area as defined in 10 CFR j
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w Part 20 and changes surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (61 FR 176). Accordingly, the amendments meet the eligibility criteria for ettegorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor:
G. P. Hornseth Date:
November 18, 1997